基于CRAM的压水堆燃耗程序开发及验证

正基于CRAM方法开发了燃耗求解程序MCRAM,利用Cinder90燃耗数据库生成了3 400阶燃耗矩阵(图1),并耦合MCNP程序对重要的锕系核素和裂变产物核素的反应截面进行了修正。以OECD/NEA乏燃料成分基准数据库中的Takahama-3压水堆燃料组件为基准题,对MCRAM程序的计算结果进行验证,并与其他程序的计算结果进行比较。结果表明,MCRAM程序对重要裂变产物和主要锕系核素的计算结果相对误差小于5%,计算精度与ORIGIN2程序的相当(图2)。     

燃耗数据库基准检验方法研究

燃耗数据库基准检验方法对于研制高准确度的燃耗数据库至关重要。本文以TAKAHAMA 3压水堆辐照后检验实验中SF95样品的建模为例，研究了建模要素对燃耗计算的影响，确定了燃耗实验建模的方法，开展了燃耗信用制研究感兴趣的锕系和裂变产物核素积存量计算值与实验值的比对。比对结果显示，主锕系核素计算偏差小于2%，大部分次锕系核素偏差小于10%，大部分重要裂变产物核素偏差小于5%。本文还对125Sb积存量随燃耗深度变化规律进行了理论分析，确认了破坏性放化实验测量结果存在缺陷，并进一步获得了125Sb积存量的修正值，使计算偏差从接近170%下降到20%以内。本次研究表明，燃耗数据库基准检验研究不仅需发展适当的燃耗实验建模方法，还需对实验数据进行适当的评价。     

蒙特卡罗燃耗计算程序MCNTRANS的开发与验证

本文介绍了开发的蒙特卡罗燃耗计算程序MCNTRANS。MCNTRANS的中子学计算参数直接采用MCNP5程序的反应率计算值,燃耗计算方法采用图论算法跟踪燃耗链,同时,对实际燃耗过程进行详细分析以提高计算精度与程序适用性,并使用预估 校正方法以获取较大的燃耗计算步长。程序计算结果通过OECD/NEA与JAERI燃耗基准题实验结果进行验证,并与其他程序的计算结果进行比较。结果表明,MCNTRANS程序在不同燃耗深度下的计算结果和实验值与其他程序的计算值符合较好,部分锕系核素与裂变产物的计算精度更高。     

基于抽样方法的燃耗计算核素积存量不确定度分析

基于抽样基本原理研究了应用于燃耗计算的不确定度分析方法,并开发了燃耗计算不确定度分析程序。基于评价核数据库ENDF/B-Ⅷ.0的裂变产额标准差和衰变常量标准差计算得到了衰变常量协方差矩阵和带相关性的裂变产额协方差矩阵,并结合SCALE6.2程序包的56群反应截面协方差数据库,对Takahama-3压水堆组件基准题中SF95-4样品进行不确定度分析。计算了反应截面、衰变常量和裂变产额不确定度引起的核素积存量的不确定度。计算结果表明,反应截面的不确定度是锕系核素积存量不确定度的主要来源,裂变产额和衰变常量的不确定度对部分裂变产物的积存量会引入较大的不确定度。但考虑裂变产额相关性后,裂变产额引起的不确定度显著降低。     

铅基快堆RBEC-M燃耗库制作方法研究及验证

      提出了一套新的方法流程，用来处理和生成燃耗计算所需的数据。利用核数据处理程序NJOY处理评价数据库ENDF-B-Ⅶ.1生成33群的MATXS格式库，再根据具体问题中的材料信息，经截面处理程序MGGC处理得到相关核素的微观、宏观截面，经自编写的处理模块Triso对其进行格式转化、合并，最终得到提供给燃耗计算程序使用的ISOTXS库文件，其中一般核素以微观截面的形式表示，裂变产物以类似宏观截面的伪裂变产物形式表示。对铅冷快堆基准题900 MW RBEC-M进行了计算，采用REBUS-3进行燃耗计算，对比了结果中的有效增殖系数k_eff随燃耗的变化趋势、功率分布以及中子能谱，最终结果与参考报告较为符合，初步验证了这一系列燃耗库制作流程的可行性。      

基于燃耗信任制的核电厂乏燃料贮存水池临界计算

为研究初始富集度为4.95%的新型燃料组件卸料后高密度贮存的可行性,以岭澳核电站3、4号机组乏燃料贮存水池为例,利用SCALE5.1程序系统中基于燃耗信任制的STARBUCS临界计算程序,分析了该新型燃料组件在不同燃耗情况下,锕系核素和裂变产物的产额变化及其对反应性的影响;基于锕系加裂变产物信任水平,计算了燃料组件在不同燃耗深度和不同贮存年限情况下的乏燃料贮存水池临界安全性;给出了乏燃料贮存水池Ⅱ区的参考装载曲线。计算表明:该新型燃料组件在燃耗达到45 GWd.t-1(U)后可以高密度贮存在乏燃料贮存水池Ⅱ区。     

基于组件计算的燃耗实验基准题建模分析

组件计算在堆芯核设计中占有重要地位。组件程序的燃耗计算精度对核反应堆堆芯的功率分布、换料寿期及反应性控制设计方面具有重要意义。为了评估用于堆芯燃耗计算的多群常数库的准确性,本文运用DRAGON计算程序建立了燃耗实验计算模型,采用SFCOMPO-2.0燃耗实验基准题提供的乏燃料样品燃耗历史参数及最终核素组分信息,对Takahama-3反应堆、H.B. Robinson-2反应堆及Beznau-1反应堆系列样品进行了建模计算,并将计算结果与SFCOMPO-2.0数据库中的基准实验结果进行了对比和分析。结果表明:多数核素的模拟结果与基准值符合良好,误差在10%以内。同时本文对理论计算值与基准实验值存在差异较大的几种核素进行了相应讨论,并对样品计算结果进行了对比分析。     

裂变产额数据对铅基快堆RBEC-M燃耗计算影响研究

第四代核能系统是一种具有更好安全性、经济竞争力、核废物减少,以及防止核扩散的先进核能系统,代表了先进核能系统的发展趋势和技术前沿。铅基快堆是第四代核能系统中重要堆型之一。目前国际上通用的反应堆程序,比如MCNP+ORIGEN、RMC或者Serpent,很多研究主要针对压水堆,国际上也有研究发现针对铅基快堆基准题RBEC-M,确定论方法和蒙卡方法计算结果有较大偏差。本文深入研究了蒙卡程序使用的裂变产额对计算结果的影响。首先对反应堆蒙特卡罗程序RMC自带和燃耗库中的部分核素的裂变产额数据进行了更新,采用国际上著名RBEC-M基准题和OECD/NEA发布的快堆Pu循环燃耗基准题进行了验证分析,计算得到了裂变份额数据对快堆燃耗计算的影响。计算结果表明:更新后的裂变产额数据对系统的有效增殖因子和主要重核的质量变化影响较小,但对部分裂变产物的质量变化影响较大,部分核素偏差超过86%。对于快堆Pu循环燃耗基准题,长寿命高放废物~(133)Cs和~(129)I的计算结果偏差分别可达22.4%和47.8%,这将对长寿命高放废物的嬗变效率和核燃料循环有重要影响。     

基于CASMO5的燃耗历史对乏燃料反应性的影响计算

基于乏燃料贮存领域常用的锕系加裂变产物(APU-2)级燃耗信任制,应用二维组件燃耗计算程序CASMO5,计算了燃耗过程中功率密度和运行历史对乏燃料k∞的影响。结果表明:燃耗计算中,选择堆芯额定功率对应的平均功率密度,同时k∞附加0.002 3的包络裕度,运行历史选择循环内及循环间无停堆额定功率运行,同时k∞附加0.004 5的包络裕度,可满足燃耗信任制中包络性原则。     

聚变-裂变混合堆程序开发及验证

针对聚变-裂变混合堆设计研究中原有燃耗计算程序MONK9A耗时长等问题,利用MCNP和SCALE5.1程序包中的Origen-s程序开发出1套可用于先进反应堆设计的燃耗耦合程序MOCouple-s.选取了压水堆燃耗基准题、ADS基准题对MOCouple-s程序进行了验证,结果表明,MOCouple-s程序关于反应性和核素成分的计算结果与实验测量结果和其他程序的计算结果符合良好,且在某些计算结果、参数设置、自动化执行等方面优于国内外类似程序.利用MOCouple-s程序对MONK9A程序在混合堆燃耗计算上的适用性进行了验证,结果差别不大,证明MONK9A程序用于混合堆初步研究设计得到的燃耗计算结果是可靠的.     

Analysis of Measured Isotopic Compositions of High-Burnup BWR MOX Fuel

Analysis of measured isotopic compositions of four high-burnup BWR MOX fuel samples was performed by using a general-purpose neutronic calculation code SRAC and a continuous-energy Monte Carlo burnup code MVP-BURN. The initial Pu fissile content of the samples was 5.52 wt%, and the burnups ranged from 50 to 80 GWd/t. It is confirmed that a geometrical model including the effect of UO₂ assemblies adjacent to the MOX assembly is necessary in the burnup calculations to obtain accurate calculated isotopic compositions. The calculated results of MVP-BURN with JENDL-3.3 taking such effect into account show more accurate results for major actinides (U, Pu, and Am isotopes) and most fission products than those of infinite assembly calculations. The paper also shows the results calculated using SRAC with JENDL-3.3, ENDF/B-VII, and JEFF-3.1.     

针对裂变产额和半衰期的燃耗计算灵敏度和不确定度分析方法

基于广义微扰理论推导了裂变产额和半衰期的燃耗灵敏度系数理论模型，该模型考虑了原子核密度和中子通量的相互影响，并开发了燃耗计算中有效增殖因数和原子核密度等响应参数对核数据的灵敏度和不确定度分析程序。基于评价核数据中裂变产物独立产额的标准差数据，产生了针对压缩燃耗数据库的裂变产额协方差矩阵，以提高不确定度的计算精度。基于ENDF/B-Ⅶ.1数据库量化了UAM基准题TMI-1栅元无限增殖因数及重要裂变产物和重核的原子核密度由裂变产额和半衰期引入的不确定度。数值结果表明，对于栅元无限增殖因数，裂变产额和半衰期引入的不确定度很小；对于部分裂变产物的原子核密度，裂变产额和半衰期会引入较大的不确定度。     

Calculation of radial burnup and nuclides atom density distributions in a VVER-1000 fuel rod using Monte Carlo method

A fundamental knowledge of fuel behavior in different situations is required for safe and economic nuclear power generation. Due to the importance of a fuel rod behavior modelling in high burnup, in this paper, the radial distribution of burnup, fission products, and actinides atom density and their variations by increasing burnup and other factors such as temperature, enrichment and power density are studied in a fuel pellet of a VVER-1000 reactor in an operational cycle using the MCNPX 2.7 Monte Carlo code. A benchmark including a Uranium-Gadolinium (UGD) fuel assembly is used for verification of the developed model in the MCNPX code for radial burnup calculation. A sensitivity study is carried out to investigate the effect of different parameters such as the number of particles per cycle, the number of geometrical radial nodes in the fuel pellet, the number of burnup steps and the selection of different fission-product contents (i.e. those isotopes that are used for particle transport) on the MCNPX model for speed and accuracy compromising. To calculate the radial temperature profiles and to analyze the effect of temperature on the radial burnup distribution and vice versa, the HEATING 7.2 code, which is a general-purpose conduction heat transfer program, and the MCNPX code are applied together. The results show the accuracy and capability of the proposed model in the MCNPX and HEATING codes for radial burnup calculation.     

Burnup Calculation Sensitivity and Uncertainty Analysis Method of Fission Yield and Decay Half-life

The calculation model of sensitivity coefficient for decay half-life and fission product yield in burnup calculation was derived based on generalized perturbation theory, which considered the interaction between nuclear concentration and neutron flux. A code was developed to calculate sensitivity and uncertainty of effective neutron multiplication factors and nuclide concentration caused by nuclear data. Covariance matrix of fission yield for a simplified burnup library was generated based on standard deviation data of independent fission yield in evaluated nuclear data library to improve the accuracy of uncertainty quantification. Uncertainties induced by decay half-life and fission yield on infinite neutron multiplication factors and nuclide concentration for TMI-1 pin-cell in the UAM burnup benchmark were quantified based on ENDF/B-Ⅶ.1. The numerical results show that the uncertainty of infinite neutron multiplication factors induced by decay half-lives and fission yields is low, while the uncertainty of concentration of some fission product nuclide is high.     

Development and validation of depletion code system IMPC Burnup for ADS

Depletion calculation is important for studying the transmutation efficiency of minor actinides and longlife fission products in accelerator-driven subcritical reactor system(ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEAADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover,the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.     

压水堆燃耗数据库的制作与验证

基于评价数据库ENDF/B-Ⅷ.0和EAF-2010研制了一套适用于CINDER90程序的压水堆用燃耗数据库,该数据库包含中子反应截面、衰变数据和裂变产额数据3部分。中子反应截面的加工分为两步,首先采用Inverted Stack算法和CRECTJ6程序将EAF 2010库的截面分支比融入ENDF/B Ⅷ0库全套中子评价数据,然后用NJOY2016程序处理成63群截面。衰变数据和裂变产额数据分别由MF8/MT457和MF8/MT454数据加工得到,裂变产额数据共包含36个裂变核的60组产额数据。以SFCOMPO 20中Takahama 3压水堆燃料组件为基准题,对研制的燃耗数据库进行了验证。结果表明,本文制作的燃耗数据库的方法是正确的,对于某些核素,如242Amm,制作的数据库比自带库的计算结果更接近实验值。